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Addendum to Hazard Summary Report Experimental Breeder Reactor-II (EBR-II)
Report containing hazard and safety information regarding the Experimental Breeder Reactor-II in Idaho.
ALPR Preliminary Design Study (Argonne Low Power Reactor) Phase 1
A preliminary design study, Phase I of the ALPR project, has been made in accordance with the Army Reactors Branch specifications for a nuclear "package" power plant with a 200-260-kw electric and 400 kw heating capacity. The plant is to be installed at the Idaho Reactor Testing Station as a prototype for remote arctic installations. The "conventional" power plant as well as the exterior reactor components are described in the accompanying report and cost estimate by Pioneer Service and Engineering Company, Architect-Engineers for the project."Nuclear" components of the reactor are designed by Argonne National Laboratory as described in the present report.
Annual Report 1961
This seventh Annual Report is a summary of some of the progress in scientific and engineering research and development carried on at Argonne National Laboratory during 1961. As is customary in this series, only those portions of the total program that have reached such a stage that they may be of general interest are recorded. Thus, a comparison with the Annual Reports for 1959 (ANL-6125) and for 1960 (ANL-6275) will reveal the description of a generally different set of scientific activities. A more detailed presentation of any work covered in this report or of the many ANL projects not mentioned may be obtained by perusing the various progress and topical reports issued by the Laboratory during 1961. A list of the publications in the scientific journals during 1961 by Argonne personnel has been given as an Appendix.
Argonne National Laboratory Chemical Engineering Division Annual Technical Report: 1980
Report of activities of Argonne Chemical Engineering Division, including advanced battery project, electro-chemical project management, advanced fuel cell development, utilization of coal, magnetohydrodynamics heat and seed recovery technology, solar energy, fast reactor chemistry research, nuclear fuel cycle studies, magnetic fusion energy research, and basic energy science.
Argonne National Laboratory Idaho Division Summary Report: July-September, 1960
Report issued by the Argonne National Laboratory covering a summary report of the work conducted by the Idaho Division. Experimental work and progress made on reactors are presented. This report includes tables, illustrations, and photographs.
Argonne National Laboratory Idaho Division Summary Report: October 1960-March 1961
Report issued by the Argonne National Laboratory discussing a summary of projects conducted with the Idaho Division. Descriptions of each project are presented. This report includes tables, and illustrations.
Chemical Engineering Division Summary Report
Measurement of radioactive carry-over was made on borax III operating at 300 psig and at power levels ranging from 4 to 14 mv. Decontamination factors of from 1.5 x 104 (at 14 mv) were obtained. These data are in essential agreement with those predicted by previous laboratory experimental work.
Chemical Engineering Division Summary Report for January, February, and March 1957
A fused fluoride process for dissolution of zirconium-uranium fuel alloys is being developed. The alloy is dissolved in an equimolar sodium fluoride-zirconium fluoride melt at 600°C by sparging the system with hydrogen fluoride. The uranium is volatilized from the melt as the hexafluoride by a sparging operation with fluorine or bromine pentafluoride vapor. This product is then decontaminated and purified by fractional distillation.
Chemical Engineering Division Summary Report for January, February, and March 1958
Development work was continued on the fused fluoride process for the recovery of enriched uranium from zirconium-matrix fuel alloys. The alloy is dissolved by immersing it in molten sodium fluoride-zirconium fluoride at 600°C and passing hydrogen fluoride vapor through the system.The dissolved uranium tetrafluoride in the melt is then volatilized as uranium hexafluoride by sparging with fluorine. The uranium hexafluoride product is purified and decontaminated by fractional distillation. Additional corrosion tests were made on a variety of metals in an effort to find a material of construction suitable for the fluorination step. All the metals tested, with the exception of Hastelloy B, were attacked rapidly in the fluorinated melt. The attack was particularly severe at the melt-gas interface when tests were made with partially submerged specimens of the metals.
Chemical Engineering Division Summary Report July, August, and September, 1956
Additional runs have been made in the six-inch, continuous-flow mixing chamber to study the rate of mass transfer between isobutanol and water. These runs were inconclusive because the effluents were mutually saturated. A new four-inch cell has been designed and is being fabricated; this will permit a reduction in the time available for mass transfer. Consideration has been given to other liquid pairs which may transfer more slowly than isobutanol-water. The system nitrobenzene-ethylene glycol appears attractive.
Chemical Engineering Division Summary Report July, August, and September, 1957
Development work continued on a fused salt process for the recovery of uranium from zirconium-matrix fuel alloys. The fuel is dissolved in a sodium fluoride-zirconium fluoride melt at 600°C by hydrogen fluoride sparging. The melt is then sparged with fluorine gas which volatilizes the dissolved uranium as the hexafluoride. The final decontamination and purification of the uranium hexafluoride are accomplished by fractional distillation. The testing of graphite as a container material for the hydrofluorination step was continued. Additional thermal cycling experiments were performed, using a helium sparge in equimolar sodium fluoride-zirconium fluoride melt at 600°C. The extent of penetration of the fused salt into the graphite was determined. No mechanical degradation was present. Dimensional change data were also obtained for graphite vessels in which the fused salt was sparged with hydrogen fluoride.
Chemical Engineering Division Summary Report October, November, and December, 1956
A final series of runs was made in a four-inch continuous-flow mixing chamber to study the transfer of isobutanol into water and nitrobenzene into ethylene glycol. Satisfactory techniques were developed to provide for the rapid analysis of these systems. In addition, a light-scattering correlation was prepared to provide a measure of the interfacial area of the yellow-colored nitrobenzene-ethylene glycol mixtures.
A Coated Cast Iron Crucible for use with Eutectic Al-Si Alloy in the Temperature Range 595°-650°C
The feasibility of the coated metal crucible as a container for eutectic Al-Si alloy has been proven by test. Small, enamel-coated cast iron pots has been proven by test. Small, enamel-coated cast iron pots have successfully withstood the chemically aggressive Al-Si alloy and the adverse influence of an oxidizing atmosphere for a period of 3 months at 725°C. A similarly coated castiron crucible containing 450 pounds of eutectic Al-Si alloy was successfully tested for 144 days in a jacketing operation conducted at 595°-650°C. Under the same conditions, the normal service life of clay-bonded graphite and silicon carbide crucibles rarely exceeds 45 days. The coating material is a commercially available enamel capable of withstanding temperatures up to 790°C (1450°F). It is readily applied to the surface of a variety of ferrous metals and alloys; however, best results are obtained with alloys low in chromium and nickel which also have a low thermal expansion coefficient.
Errata for ANL-6628: Automatic Foil Activity Counting Facility and Data-Reduction Program
Errata sheet listing corrections to three pages of a report that describes a transistorized automatic counting and recording system built for the determination of foil-activation data.
The Fabrication of a Plutonium Helix for a Doppler Experiment
A helix constructed of plutonium was made to test the Doppler temperature effect in ZPR-III. The helix, 1 inch in diameter and 6-1/4 inches long, contained 240 grams of delta-phase plutonium alloy encapsulated in titanium tubing. Four plutonium rods were extruded, joined together, and pushed into a titanium tube. This tube was swaged tightly over the plutonium rod, and the assembly was wound into a coil. Electrical leads to the coil were made by swaging copper tubing over the ends of the coil. The helix was tested by cycling about 500 times between 50°C and 190°C. The coil was heated with a current of 130 amperes and cooled with a blast of chilled helium. (1) Several helices of uranium(2) were cycled during the same tests. Despite the severity of the thermal cycles, the helices were undamaged.
The Fabrication of Prototype Fuel Elements for the Experimental Boiling Water Reactor and the Experimental Breeder Reactor
The purpose of this program was to develop techniques and methods for producing fuel elements for the Experimental Boiling Water and Experimental Breeder Reactors. Methods for fabricating large tubes, flat plates, and small pins were investigated. The tube and plates contained U-5 w/o Zr-1.5 w/o Nb alloy and were designed for the EBWR. The pins contained U-2 w/o Zr alloy and were designed for the EBR. Cladding and end seal material of Zircaloy-2 was required for the water-cooled EBWR elements. Unalloyed zirconium was specified for cladding on the sodium-cooled EBR elements.
Interim Report: Faret Experimental Program
From Introduction: "The Fast Reactor Test (FARET) program is part of the United States Fast Reactor Program directed toward the demonstration of breeder reactors essential to the long-range goal of utilizing fertile fuels. (1) With-in this program is the current thinking that the early construction and operation of two or three prototype nuclear power plants of the order of 250 Mw(e) will lead eventually to the construction of practical and economic full scale breeder reactors by the early 1980's. (2)"
A laboratory Ivestigation of the Fluorination of Crude Uranium Tertrafluoride
Ore concentrates have been converted directly to crude uranium tetrafluoride by hydrogen reduction and hydrofluorination in fluidized-bed reactors. Small-scale laboratory experiments demonstrated that this process can be extended to the production of crude uranium hexafluoride through fluorination of the uranium tetrafluoride in a fluidized bed. The satisfactory temperature range for the reaction lies between 300°C and 600°C. At 450°C the fluorine utilization is between 50 and 80 per cent. With excess fluorine, over 99 per cent of the uranium is volatilized from the solid material. The fluidization characteristics of certain materials are improved by the addition of an inert solid diluent to the bed.
Land Reclamation Program, Annual Report: July 1975-July 1976
Annual report of the Argonne National Laboratory Land Reclamation Program outlining the activities and research conducted during the year, various assessments and discussion, and related documentation.
The Manufacture of Enriched ZPR-III Fuel Plates
This report is essentially a procedural account of the fabrication of certain enriched ZPR-III fuel plates for use in the ANL fast critical experiments at Arco, Idaho. A total of 208.92 kilograms of fully enrich, unalloyed uranium was processed. Of this amount 202.74 kilograms was received in the form of Oak Ridge type reduction buttons and 6.18 kilograms as pressed-powder plates. The completed fabrication consisted of 720 rectangular fuel plates having the nominal dimensions 3in. x 2in. x 1/8in. Their combined weight of 159.21 kilograms represents 76.22% of the weight of enriched material processed. The final distribution of the enriched material was as follows: [figure not transcribed].
Metallurgy Division Quarterly Report [for] October, November, and December 1955
A total of nine clad plates, containing uranium -5 w/o zirconium 1.5 w/o niobium alloy cores and clad with Zircaloy-II, were rolled in plain carbon steel jackets, heat treated, physically evaluated, and corrosion tested. All these plates were found to be within predetermined dimensional tolerance in width, thickness, length, cladding thickness, and core distribution. Improved control of wielding variables and of the length of the seal pin projecting above the end plugs resulted in the elimination of frequently observed segmented inclusions at the seal pin interfaces.
A Multiple-Source Urban Atmospheric Dispersion Model
Report documenting the development phase of a multiple-source, urban atmospheric dispersion model that describes environmental transients.
Operating Manual for the Argonaut Reactor
The design of the Argonaut (Argonne Nuclear Assembly for University Training) was initiated by the Reactor Engineering Division of Argonne National Laboratory to satisfy needs for a low-power reactor facility within the Laboratory, and for training uses within the international School of Nuclear Science and Engineering (ISNSE). It was intended primarily for instruction and research in reactor physics. It was also considered as a possibility that it would fulfill the requirements of universities engaged in a program of nuclear science. The cost of the facility was to be kept to a minimum consistent with the high degree of inherent safety and a great amount of flexibility in the system. The basic design stemmed from the Knolls Atomic Power Laboratory Thermal Test Reactor* (TTR), now called Nuclear Test Reactor (NTR). Modification during the course of the work justified the new name "Argonaut".
Preliminary Design Requirements Argonne Boiling Reactor (ARBOR) Facility
From Introduction: "Descriptions of the functional requirements of the facility, together with preliminary concepts of methods for meeting them, are presented in this prospectus."
Proceedings of the 1978 Symposium on Instrumentation and Control for Fossil Demonstration Plants : June 19-21, 1978, Newport Beach Marriott, Newport Beach, California
This conference covers various aspects of fossil-fuel power plants based on coal or coal-conversion products, as well as the process control equipment involved in the conversion or combustion processes.
Quarterly Progress Report on Reactor Development 400 Program
Physics calculations have been made for various combinations of the four types of fuel assemblies to be used in the EBWR core. Two thicknesses of plates, 0.205 in. and 0.274 in., including the two 0.020-in. cladding layers, are to be made of both natural U and U containing 1.44% U235. A total of 148 assemblies, 74 natural and 74 enriched, are to be fabricated with six identical plates each. Various configurations of these fuel assemblies will be used to (1) change the critical size of the core, (2) change the power distribution in the core, and (3) change the amount of reactivity corresponding to a given stream volume in the core. The physics calculations show that uncertainties in critical mass are adequately covered by the number and variety of fuel assemblies and that the possible changes in core characteristics with the different fuel assemblies should provide valuable information about the factors affecting maximum power density and stability in a boiling water reactor.
Quarterly Report January, February and March, 1956
The EBWR loading requires a total of 888 plates. It is anticipated that approximately 1000 plates will have to be produced to obtain the number of acceptable plates required for the loading. To the end of this quarter, 568 cladding billet cores acceptable with respect to chemical composition and physical soundness had been cast; this number represents 78% of the total number of cores cast. Approximately 75% of the Zircaloy-II stock required has been rolled, and about 55% of the cladding components required have been finished. The anticipated number of 495 cladding billets required for the thin (0.210") natural and enriched plates have been assembled, welded, sealed, and jacketed in steel. A total of 310 cladding billets have been rolled to fuel plates; of this number, 142 have been completely finished, and the remaining 168 are in the finish processing stages. The stability of the equipment for measuring the clad thickness of EBWR fuel plates has been improved by placing the phototube and the anthracene scintillator crystals in an insulated box with a temperature regulation of the order of 0.1°F.
Quarterly Report October, November and December, 1956
Methods of producing extremely clean surfaces on rolled Zircaloy-2 strip have been investigated. It has been found that the finer abrasives, 400 mesh or finer, are more effective than coarse types because of their ability to penetrate pits and crevices more readily. Two such cleanings, with an intermediate 35 v/o HNO3-5 v/o HF pickle, resulted in a microscopically clean surface. Ultrasonic inspection of the EBWR fuel plates has been completed during this quarter. Approximately 95% of the plates were found acceptable. All subassemblies manufactured from the EBWR plates met dimensional specifications and passed 9-day corrosion tests at 290°C (550°F). All thoria-urania pellets for the loading of Borax-IV have been pressed, loaded into tube plates, and fabricated into subassemblies. The total number of subassemblies made was 82, of which 72 were fuel plates and 10 were blanket plates, more than sufficient for the loading. The reactor has gone critical using this loading.
Reactor Engineering Division Quarterly Report [for] October, November, December 1955. Section I
The gastight steel building (400,000 cu ft) in which all radioactive components are to be housed has been completed by the Graver Tank Company. This structure was tested for strength at 18.75 psig (20% above design pressure) and then tested for leaks. No leaks were found in soap bubble testing of all welded seams. Continuous measurements of temperature and pressure over a ten-day period showed the leakage, if any, to be less than the 500 cu/ ft/day at 15 psig specified. The gastight cylinder was, therefore, accepted. General construction work by the Sumner Sollitt Company on the remainder of the plant has begun.
Reactor Engineering Division Quarterly Report Section I January, February, March. 1956
Physical calculations have been performed for various combinations of the four types of fuel assemblies to be used in the EBWR core. Two thicknesses of plates (0.205 in. and 0.274 in., including two 0.020-in. cladding layers) are to be made of both natural uranium and uranium containing 1.44% U235. Any given fuel assembly contains six identical plates. A total of 148 assemblies, 74 natural and 74 enriched, are to be fabricated. Various configurations of these fuel assemblies can be used to (1) change the critical size of the core, (2) change the power distribution in the core or (3) change the amount of reactivity corresponding to a given steam volume in the core. Physics calculations show that any uncertainties in the required critical mass are adequately covered by the number and variety of fuel assemblies, and that the changes in core characteristics possible with the different fuel assemblies should provide valuable information about the factors affecting maximum power density and stability in a boiling reactor.
Stirling-Engine Thermodynamic Analysis: A Users Guide To SEAM1
From Abstract: This report provides background and procedural information for the use of a general-purpose Stirling-engine analysis code developed at Argonne National Laboratory and available through the National Energy Software Center. Different engine configurations are easily specified, or the user may make use of provided data for existing engines, both kinematic and free piston. The code models heat transfer and fluid mechanics throughout the engine and accounts for system energy flows and losses. Good agreement is shown between code predictions and experimental measurements. The present analysis method was chosen for fast execution and useful information on energy flows in the system. A description is provided of the code structure that is intended to accept different analysis modules that can provide improved system modeling or optimization capability.
Summary of Conference on the Toxicity of Carbon 14 Held at Argonne National Laboratory, January 15-16, 1952
Report issued by the Argonne National Laboratory discussing the Conference on the Toxicity of Carbon 14. At the outset of the meeting it was emphasized that the group was brought together to discuss work which had a bearing on the toxicity of C14 compounds used either clinically or industrially, and the hope was expressed that the group might be able to reach some conclusions on these matters. This report includes tables.
Summary of Meteorological Data Taken at Argonne National Laboratory, Du Page County, Illinois, July 1952 Through June 1953
Report issued by the Argonne National Laboratory discussing meteorological data collected between 1952 and 1953. Wind, temperature, pressure, and precipitation studies are presented. This report includes tables, illustrations, and a map.
Summary Report of the Hazards of the Internal Exponential Experiment (ZPR-V)
The Internal exponential Exponential Experiment (ZPR-V) will be constructed by loading up to 49 of the fuel cans, containing up to 155 kg of U235, of the present Fast Exponential Experiment in a 22-in. square iron tank, surrounded by an annular thermal region of fully enriched light water lattice 10 to 15 cm thick. This assembly will be placed in a 5-ft diameter tank which will, in turn, be located in the 10-ft diameter ZPR-II tank, the annular space between the outer tanks containing water for shielding. The new experiment will be a well-shielded, strongly coupled fast-thermal system. It will be possible to make measurements that cannot be made on the present Fast Exponential Experiment. One category of such determinations is the study of reactivity effects produced in the fast core, including control scheme studies and danger coefficient and oscillator measurements of such effects as Doppler coefficients and effect of lumping and streaming. The higher flux and excellent shielding will make beam studies of energy spectrum practical. Additional foil activations will be possible. Characteristics of mixed fast-thermal systems, which are of potential importance as power breeders, can be studied.
Table of Sin θ and Sin2 θ for Values of θ from 2° to 87°
The table of sin θ and sin2 θ, to five decimal places for every hundreth of a degree from 2°-87°, has been prepared for the use of Professor W. H. Zachariasen in his X-ray diffraction studies. [Tables not transcribed]
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